Protection systems for nuclear boiling water reactors

ABSTRACT

A protection system for a nuclear boiling water reactor may include a device configured to monitor reactor power; a device configured to monitor reactor pressure; a device configured to determine a power-dependent high reactor pressure setpoint, based on the monitored reactor power; and a device configured to initiate a protection system action when the monitored reactor pressure is greater than the power-dependent high reactor pressure setpoint. The power-dependent high reactor pressure setpoint that corresponds to at least one value of percent power in an operating domain of the reactor may be less than the power-dependent high reactor pressure setpoint that corresponds to 100% reactor power.

PRIORITY STATEMENT

This application is a divisional application of U.S. patent applicationSer. No. 11/642,920, filed on Dec. 21, 2006, and claims the associatedbenefit under 35 U.S.C. §121. The entire contents of parent U.S. patentapplication Ser. No. 11/642,920, entitled “PROTECTION SYSTEMS FOR ANDMETHODS OF OPERATING NUCLEAR BOILING WATER REACTORS”, are incorporatedherein by reference.

BACKGROUND

1. Field

Example embodiments relate to protection systems for and methods ofoperating nuclear boiling water reactor (“BWR”) power plants.

2. Description of Related Art

FIG. 1 illustrates a related art BWR. As shown, a pump 100 supplieswater to a reactor vessel 102 housed within a containment vessel 104.The core 106 of the reactor vessel 102 includes a number of fuel bundlessuch as those described in detail below with respect to FIG. 2. Thecontrolled nuclear fission taking place at the fuel bundles in the core106 generates heat that turns the supplied water into steam. This steamis supplied from the reactor vessel 102 to turbines 108 that power agenerator 110. The generator 110 then outputs electrical energy. Thesteam supplied to the turbines 108 is recycled by condensing the steamfrom turbines 108 back into water at a condenser 112, and supplying thecondensed steam back to the pump 100.

FIG. 2 illustrates a typical fuel bundle 114 in the core 106. A core 106may include, for example, anywhere from about 200 to about 900 of thesefuel bundles 114. As shown in FIG. 2, the fuel bundle 114 may include anouter channel 116 surrounding a plurality of fuel rods 118 extendinggenerally parallel to one another between upper and lower tie plates 120and 122, respectively, and in a generally rectilinear matrix of fuelrods as illustrated in FIG. 3, which is a schematic representation of across-section or lattice of the fuel bundle 114 of FIG. 2. The fuel rods118 may be maintained laterally spaced from one another by a pluralityof spacers 124 vertically spaced apart from each other along the lengthof the fuel rods 118 within the outer channel 116. Referring to FIG. 3,there is illustrated in an array of fuel rods 118 (i.e., in thisinstance, a 10×10 array) surrounded by the outer channel 116. The fuelrods 118 are arranged in orthogonally related rows and also surround oneor more “water rods,” two water rods 126 being illustrated. The fuelbundle 114 may be arranged, for example, in one quadrant of a controlblade 128 (also known as a “control rod”). It will be appreciated thatother fuel bundles 114 may be arranged in each of the other quadrants ofthe control blade 128. Movement of the control blade 128 up and/or downbetween the fuel bundles 114 controls the amount of reactivity occurringin the fuel bundles 114 associated with that control blade 128.

The total number of control blades 128 utilized varies with core sizeand geometry, and may be, for example, between about 50 and about 200.The axial position of the control blades 128 (i.e., fully inserted,fully withdrawn, or somewhere in between) is based on the need tocontrol excess reactivity and to meet other operational constraints. Foreach control blade 128, there may be, for example, 24, 48, or morepossible axial positions or “notches.”

The BWR may include several related art closed-loop control systems thatcontrol various individual operations of the BWR in response to demands.For example, a related art recirculation flow control system (“RFCS”)may be used to control core flowrate that, in turn, help to determinethe output power of the reactor core. A control blade drive systemaffects the position of the control blades, the control blade densitywithin the core, and core reactivity. A turbine control system controlssteam flow from the BWR to the turbines based on load demands andpressure regulation.

The operation of all of these systems, as well as other related artsystems, is controlled utilizing various monitoring parameters of theBWR. Exemplary monitoring parameters include core flow and flowrateeffected by the RFCS, reactor vessel dome pressure (which is thepressure of the steam discharged from the pressure vessel to theturbines), neutron flux or core power, feedwater temperature andflowrate, steam flowrate provided to the turbines, and various statusindications of the BWR systems. Many monitoring parameters are measureddirectly by related art sensors, while others, such as core thermalpower, are typically calculated using measured parameters. These statusmonitoring parameters are provided as output signals from the respectivesystems.

Nuclear reactors are conservatively specified to minimize any risks fromthe hazardous materials involved in their use. The materials used inBWRs must withstand various loading, environmental, and radiationconditions. For example, operating pressures and temperatures for thereactor pressure vessel are about 7 MPa and 290° C. for a BWR. Reactorvessel walls are thus several inches thick and very strong materials areused for reactor components. Nonetheless, contingencies are required forfailure as components are subjected to operational stress for decades.These contingencies involve not only many layers of preventive systems,but also procedures for rectifying problems that arise.

Related art reactor control systems have automatic and manual controlsto maintain safe operating conditions as the demand is varied. Theseveral control systems control operation of the reactor in response togiven demand signals. Computer programs are used to analyze thermal andhydraulic characteristics of the reactor core. The analysis is based onnuclear data selected from analytical and empirical transients andaccidents, and from reactor physics and thermal-hydraulic principles. Inthe event of an abnormal transient, the reactor operator usually is ableto diagnose the situation and take corrective action based on applicabletraining, experience, and/or judgment. Whether the manual remedialaction is sufficient depends upon the transient and upon the operator'sknowledge and/or training. If the transient is significant (i.e.,challenges any of the reactor safety limits), reactor trip (alsoreferred to as reactor shutdown, scram, or full insertion of all controlblades) may be required (the term “scram” is alleged to have originatedin the early years of reactor development and operation as an acronymfor “super-critical reactor ax-man”). Some transients may occur quickly(i.e., faster than the capability of a human operator to react). In sucha transient, a reactor trip will be initiated automatically. Safetyanalyses generally show that no operator action is necessary within 10minutes of a postulated transient.

A related art nuclear reactor protection system (“RPS”) comprises amulti-channel electrical alarm and actuating system that monitorsoperation of the reactor and, upon sensing an abnormal transient,initiates action to prevent an unsafe or potentially unsafe condition.At minimum, the related art RPS typically provides three functions: (1)reactor trip that shuts down the reactor when certain monitoredparameter limits are exceeded; (2) nuclear system isolation thatisolates the reactor vessel and all connections penetrating acontainment barrier; and (3) engineered safety feature actuation thatactuates related art emergency systems such as cooling systems andresidual heat removal systems.

Core power protection schemes are typically employed in BWRs when thereactor is operating in its normal operating domain (i.e., after startupand heatup of the reactor). FIG. 4 is a typical BWR power-to-flowoperating map showing an operating domain of the reactor. Such operatingdomains are discussed, for example, in U.S. Pat. No. 5,528,639 (“the'639 patent”). After startup and heatup, the permissible operatingdomain for the BWR typically is above the cavitation region, below themaximum operating line, and bounded by the minimum normal flow line andthe maximum normal flow line. In related art RPSs, when the BWR isoperating within the operating domain, an unplanned transient that doesnot increase the power level (i.e., neutron flux) above a setpoint orsetpoints associated with the maximum operating line will not cause areactor overpower protection trip. FIG. 5 is a BWR power-to-flowoperating map showing an operating domain of the reactor with expandedlimits. Such operating domains are discussed, for example, in U.S. Pat.Nos. 6,721,383 B2 (“the '383 patent”) and 6,987,826 B2 (“the '826patent”). FIG. 6 is a BWR power-to-flow operating map showing anotheroperating domain of the reactor with expanded limits. Such operatingdomains are also discussed, for example, in the '383 patent and the '826patent. The disclosures of the '639 patent, the '383 patent, and the'826 patent are incorporate in this application by reference.

A reactor overpower protection trip is initiated for certain transientsthat could cause an increase in power above the maximum safe operatinglevel. Generally, an overpower equal to about 120% of the rated powercan be tolerated without causing damage to the fuel rods. If thermalpower should exceed this limiting value (the maximum safe operatinglevel) or if other abnormal conditions should arise to endanger thesystem, the RPS will cause a reactor trip.

An essential requirement of an RPS is that it must not fail when needed.Therefore, unless the operator promptly and properly identifies thecause of an abnormal transient in the operation of the reactor, andpromptly effects remedial or mitigating action, related art RPS willautomatically effect reactor trip. However, it is also essential thatreactor trip be avoided when it is not desired or necessary (i.e., whenthere is an error in the instrumentation or when the malfunction issmall enough that reactor trip is unnecessary).

As discussed in U.S. Pat. No. 5,528,639 (“the '639 patent”), forexample, four power-related methods may be used to ensure thatacceptable fuel and reactor protection are maintained. Each method usesmonitored neutron flux to sense when an increase in power occurs, buteach employs a different approach to initiate reactor trip.

The first method of protection causes a reactor overpower protectiontrip if the monitored neutron flux exceeds a preselected and fixed firstsetpoint. This first setpoint may be, for example, about 120%-125% ofrated power.

The second method of protection causes a reactor overpower protectiontrip if the monitored neutron flux exceeds a preselected, butflow-referenced, second setpoint. In this method, the second setpoint isequal to the first setpoint when the reactor core flow is high. However,when reactor core flow is reduced, the second setpoint is also reduced.

The third method of protection involves electronically filtering themeasured neutron flux signal to produce a signal that has been calledsimulated thermal power (“STP”). Usual practice employs asingle-time-constant filter that approximates the thermal response ofthe reactor fuel rods. A reactor overpower protection trip is initiatedwhen the STP signal exceeds the flow-referenced second setpoint. Thissecond setpoint may be, for example, about 110%-115% of rated power. Thethird method is usually used in combination with the first method.

In the three methods discussed above, the reactor overpower protectiontrip setpoints are above the normal operating domain of the reactor toavoid undesired trips during operation in the upper portion of theoperating domain. If more protection is required due to partial corepower and flow conditions, the reactor overpower protection tripsetpoints are manually adjusted. These manual adjustments are acumbersome nuisance for reactor operators. However, if the reactoroverpower protection trip setpoints are not adjusted, complex andrestrictive core operating limits are required to ensure acceptableprotection at all operating power and flow conditions.

Slow transients have been postulated in the partial power and flow rangethat challenge the effectiveness of these three related art protectionmethods. These slow transients have been postulated to avoid theprotection provided by the associated reactor overpower protection tripsetpoints.

As discussed in the '639 patent, a fourth method of protection involvesautomatically adjusting reactor overpower protection trip setpoints tobe a controlled margin above the operating power level of the BWR. Thefourth method provides enhanced reactor protection when the reactor isoperating at less than the maximum operating level. However, alternateand/or supplemental methods of protection may be desired.

SUMMARY

Example embodiments may provide protection systems for operating nuclearBWR power plants. Also, example embodiments may provide methods ofoperating nuclear BWR power plants.

In an example embodiment, a protection system for a nuclear BWR mayinclude a power-dependent high reactor pressure setpoint. The highreactor pressure setpoint that corresponds to at least one value ofpercent power in an operating domain of the reactor may be less than thehigh reactor pressure setpoint that corresponds to 100% power.

In another example embodiment, a protection system for a nuclear BWR mayinclude a first high reactor pressure setpoint that corresponds to 100%power and at least one second high reactor pressure setpoint thatcorresponds to one or more values of percent power in an operatingdomain of the reactor. The at least one second high reactor pressuresetpoint may be less than the first high reactor pressure setpoint.

In yet another example embodiment, a method of operating a nuclear BWRmay include implementing, in a protection system for the reactor, apower-dependent high reactor pressure setpoint. The high reactorpressure setpoint that corresponds to at least one value of percentpower in an operating domain of the reactor may be less than the highreactor pressure setpoint that corresponds to 100% power.

In a further example embodiment, a method of operating a nuclear BWR mayinclude implementing, in a protection system for the reactor, a firsthigh reactor pressure setpoint that corresponds to 100% power, andimplementing, in the protection system for the reactor, at least onesecond high reactor pressure setpoint that corresponds to one or morevalues of percent power in an operating domain of the reactor. The atleast one second high reactor pressure setpoint may be less than thefirst high reactor pressure setpoint.

BRIEF DESCRIPTION OF THE DRAWINGS

The present invention will become more fully understood from thedetailed description given below and the accompanying drawings, whereinlike elements are represented by like reference numerals, which aregiven by way of illustration only and thus are not limiting on thepresent invention and wherein:

FIG. 1 illustrates a related art BWR;

FIG. 2 illustrates a fuel bundle in the core of FIG. 1;

FIG. 3 is a schematic representation of a cross-section or lattice ofthe fuel bundle of FIG. 2;

FIG. 4 is a typical BWR power-to-flow operating map showing an operatingdomain of the reactor;

FIG. 5 is a BWR power-to-flow operating map showing an operating domainof the reactor with expanded limits;

FIG. 6 is a BWR power-to-flow operating map showing another operatingdomain of the reactor with expanded limits;

FIG. 7 is a graph showing a typical curve of reactor pressure versusreactor power for a BWR;

FIG. 8 is a graph showing a typical curve of reactor pressure versusreactor power for a BWR, a related art pressure setpoint line, and anexample new pressure setpoint curve;

FIG. 9 is a graph showing a typical curve of reactor pressure versusreactor power for a BWR, a related art pressure setpoint line, and twoexample new pressure setpoint curves;

FIG. 10 is a graph showing a typical curve of reactor pressure versusreactor power for a BWR, a related art pressure setpoint line, and anexample new pressure setpoint curve;

FIG. 11 is a graph showing a typical curve of reactor pressure versusreactor power for a BWR, a related art pressure setpoint line, and anexample new pressure setpoint curve; and

FIG. 12 illustrates a protection system for a nuclear boiling waterreactor according to example embodiments.

DETAILED DESCRIPTION OF EXAMPLE EMBODIMENTS

Example embodiments will now be described more fully with reference tothe accompanying drawings. Embodiments, however, may be embodied in manydifferent forms and should not be construed as being limited to theexample embodiments set forth herein. Rather, these example embodimentsare provided so that this disclosure will be thorough and complete, andwill fully convey the scope to those skilled in the art.

As used herein, the term “and/or” includes any and all combinations ofone or more of the associated listed items.

It will be understood that, although the terms first, second, third,etc., may be used herein to describe various elements, components,regions, layers, and/or sections, these elements, components, regions,layers, and/or sections should not be limited by these terms. Theseterms are only used to distinguish one element, component, region,layer, or section from another element, component, region, layer, orsection. Thus, a first element, component, region, layer, or sectiondiscussed below could be termed a second element, component, region,layer, or section without departing from the teachings of the exampleembodiments.

The terminology used herein is for the purpose of describing particularexample embodiments only and is not intended to be limiting. As usedherein, the singular forms “a,” “an,” and “the” are intended to includethe plural forms as well, unless the context clearly indicatesotherwise. It will be further understood that the terms “comprises,”“comprising,” “includes,” and/or “including,” when used in thisspecification, specify the presence of stated features, integers, steps,operations, elements, and/or components, but do not preclude thepresence or addition of one or more other features, integers, steps,operations, elements, and/or components.

Unless otherwise defined, all terms (including technical and scientificterms) used herein have the same meaning as commonly understood by oneof ordinary skill in the art to which example embodiments belong. Itwill be further understood that terms, such as those defined in commonlyused dictionaries, should be interpreted as having a meaning that isconsistent with their meaning in the context of the relevant art andshould not be interpreted in an idealized or overly formal sense unlessexpressly so defined herein.

Reference will now be made to example embodiments, which are illustratedin the accompanying drawings, wherein like reference numerals refer tothe like components throughout.

As discussed above, a RPS may monitor operation of the reactor. Itemsmonitored may include, for example, main steam isolation valveposition(s), turbine stop valve position(s), fast closure of turbinecontrol valve(s), drywell pressure, reactor dome pressure (referred tobelow as “reactor pressure”), reactor water level, main steam lineradiation, and/or reactor neutron flux—possibly including STP—referredto below as “reactor power.”

One or more warnings, alarms, and/or mitigating actions may beinitiated, for example, in the event of abnormal main steam isolationvalve position(s), turbine stop valve position(s), fast closure ofturbine control valve(s), high drywell pressure, high reactor pressure,low reactor water level, main steam line high radiation, and/or reactoroverpower. In addition or in the alternative, a reactor scram may beinitiated, for example, in the event of abnormal main steam isolationvalve position(s), turbine stop valve position(s), fast closure ofturbine control valve(s), high drywell pressure, high reactor pressure,low reactor water level, main steam line high radiation, and/or reactoroverpower. The reactor scram may be initiated, for example, by a lonescram signal or by more than one scram signal (possibly includingredundancy and/or “voting,” both known to one of ordinary skill in theart). The scram signal or signals may be routed via a common scram bus.

RPS monitoring of valve closure(s) related to main steam isolation valveposition(s), turbine stop valve position(s), and/or fast closure ofturbine control valve(s) can be complex because a reactor plant mayinclude, for example, eight or more main steam isolation valves, four ormore turbine stop valves, and/or four or more turbine control valves.Other issues for RPS monitoring of valve closure(s) may include:

(1) below about 25% reactor power, for example, reactor plants may notbe required to monitor thermal limits;

(2) below about 30%-40% reactor power, for example, a low-power bypass(commonly referred to as “P_(bypass)”) or equivalent setting may disablescrams due to valve closure(s) because steam flow from the reactor islow enough to be bypassed directly to the condenser, obviating the needfor a scram (use of and settings for such a low-power bypass orequivalent setting may be plant-dependent);

(3) below about 40%-60% reactor power, for example, power-to-loadunbalancing (“PLU”) may prevent fast closure—typically on the order ofabout 100 milliseconds—of the turbine control valves, allowing only slowclosure—typically on the order of about 5 seconds—that may limit and/orprevent turbine overspeed when the slow closure does not generate adirect scram signal (use of and settings for PLU may beplant-dependent);

(4) maintenance may affect the monitoring of valve closure(s) due to therespective valve(s) being out of normal operating position; and/or

(5) other plant-specific needs and design features.

As a result, some reactor plant transients may not result in a directscram due to valve closure(s). However, in a BWR, such valve closure(s)may result in transients that raise both reactor pressure (i.e., loss ofsteam demand may cause reactor pressure to rise) and reactor power(i.e., the higher reactor pressure may collapse some voids in the core,adding positive reactivity).

Although a high reactor pressure scram typically may be a backup scramto a valve closure(s) scram and/or a reactor overpower scram, there maybe situations in which the high reactor pressure scram should be theprimary scram mechanism. Pressurization transients, for example, thatmay require a scram for reactor protection—but that do not have a directscram on valve closure(s)—may scram on reactor overpower and/or highreactor pressure. However, the time required to reach a scram on reactoroverpower starting from a reduced level of reactor power may be longer,due to the larger difference between the reduced level of reactor powerand the setpoint for scram on reactor overpower. As a result, a scram onhigh reactor pressure may need to be relied upon. At the same time, thereactor pressure associated with the reduced level of reactor power maybe lower as well. As a result, the time required to reach a scram onhigh reactor pressure starting from a reduced level of reactor power mayalso be longer, due to the larger difference between the reduced reactorpressure and the setpoint for scram on high reactor pressure.

Related art solutions to this problem may involve calculating morerigorous thermal limits for reactor operation at reduced levels ofreactor power. However, these more rigorous thermal limits—oftenreferred to as thermal limit “penalties”—may complicate reactoroperation at reduced levels of reactor power, may severely limitmaneuverability of the reactor at the reduced levels of reactor power,and/or may increase reactor power ascension times.

FIG. 7 is a graph showing a typical curve of reactor pressure (psia)versus reactor power (% of rated power) for a BWR. As may be seen,reactor pressure tends to go up as reactor power goes up, but therelationship is not linear. Although the typical curve of FIG. 7 extendsonly from about 25% reactor power to about 100% reactor power, thetypical curve would extend down and to the left and/or up and to theright according to the calculations discussed below.

The operating pressure of the reactor at given values of reactor power,or “operating pressure,” may be calculated approximately as:

operating pressure=control value+P/P ₀*PCB+(P/P ₀)²*SLPD

where:

the control value at rated conditions (“control value”)—which isindependent of reactor power—is a pressure setting in the turbinecontrol system calculated to achieve rated steam pressure at rated steamflow;

P represents actual reactor power;

P₀ represents rated reactor power;

PCB represents the pressure control band; and

SLPD represents the steam line pressure drop.

PCB may be calculated as:

PCB=turbine inlet pressure−control value

The turbines may be controlled, for example, by a turbine controlsystem. The turbine control system may use turbine inlet pressure as apressure it tries to maintain. Maintaining the turbine inlet pressureconstant may result, for example, in lower reactor pressures at lowsteam flow rates. The turbine inlet pressure may be measured, forexample, in a common steam header upstream of the turbine stop valvesand the turbine control valves.

PCB may be converted to percent steam flow by multiplying the PCBpressure value by the pressure regulator gain. Because the PCB may havea range of about 0 psi-30 psi and because the pressure regulator gainmay be approximately linear over the full range of rated steam flow(0%-100%), the pressure regulator gain may be approximately constant ata value of about 3.33% steam flow per psi. Thus, at rated conditions—30psi PCB—the control system produces 100% steam flow demand.

SLPD may be calculated as:

SLPD=operating pressure−turbine inlet pressure

Values for SLPD, for example, may be greater than or equal to about 30psi and less than or equal to about 100 psi. In an example embodiment,the SLPD may be greater than or equal to about 55 psi and less than orequal to about 70 psi. In another example embodiment, the SLPD may beabout 65 psi.

Because of the dependence of operating pressure on, for example, thecontrol value, rated reactor power, and turbine inlet pressure, thevalues of operating pressure may be plant-dependent. Values foroperating pressure corresponding to 100% power, for example, may begreater than or equal to about 1,000 psia and less than or equal toabout 1,075 psia. In FIG. 7, the operating pressure corresponding to100% power is about 1,050 psia.

Steam flow may not be exactly proportional to reactor power due, inpart, to changes in feedwater temperature (water supplied from the pump100 to the reactor vessel 102) as reactor power changes. A more exactcalculation may substitute the mass flow rate of steam divided by therated mass flow rate of steam for the P/P₀ term. The difference betweenthe two calculations may be on the order of 2%-5%, with the approximatecalculation of operating pressure generally being closer to the moreexact calculation of operating pressure as reactor power increases.

FIGS. 8-11 are graphs showing typical curves of reactor pressure (psia)versus reactor power (% of rated power) for a BWR, related art pressuresetpoint lines, and examples of a new pressure setpoint curve (“NPSC”)or NPSCs according to some of the example embodiments. Although thetypical curves, setpoint lines, and NPSCs of FIGS. 8-11 extend only fromabout 25% reactor power to about 100% reactor power, the typical curves,setpoint lines, and/or NPSCs could be continued to lower and/or highervalues of reactor power. Generally, the NPSC at higher values of reactorpower may not be higher than the high reactor pressure setpoint thatcorresponds to 100% power.

FIG. 8 is a graph showing a typical curve of reactor pressure (psia)versus reactor power (% of rated power) for a BWR, a related artpressure setpoint line, and an example NPSC that may include zero-order,first-order, and/or second-order components.

Related art BWR RPSs use a fixed high reactor pressure setpoint for alllevels of reactor power. The fixed high reactor pressure setpoint maybe, for example, determined based upon a reactor pressure correspondingto 100% power. As discussed above, values for reactor pressurecorresponding to 100% power, for example, may be greater than or equalto about 1,000 psia and less than or equal to about 1,075 psia. In anexample embodiment, the reactor pressure corresponding to 100% power maybe about 1,050 psia. At least partially as a result, values for thefixed high reactor pressure setpoint in related art reactors, forexample, may be greater than or equal to about 1,040 psia and less thanor equal to about 1,125 psia. In FIG. 8, the fixed high reactor pressuresetpoint is about 1,115 psia. If the setpoint line was continued tolower and/or higher values of reactor power, the setpoint at those lowerand/or higher values of reactor power also would be about 1,115 psia.

In an example embodiment, a protection system for a nuclear BWR mayinclude a power-dependent high reactor pressure setpoint (“PDHRPS”). ThePDHRPS may take on two or more values in an operating domain of thereactor, depending on the value of power that is, for example, measured,calculated, or measured and calculated. Values of power not expressed inpercent power may be so expressed by dividing, for example, themeasured, calculated, or measured and calculated power by the ratedpower. The high reactor pressure setpoint that corresponds to at leastone value of percent power in the operating domain of the reactor may beless than the high reactor pressure setpoint that corresponds to 100%power.

The example PDHRPS/NPSC of FIG. 8 maintains a substantially constantpressure difference between itself and the operating pressure curve. ThePDHRPS may be calculated as:

PDHRPS=control value+P/P ₀*PCB+(P/P ₀)²*SLPD+HRPS₁₀₀−RP₁₀₀

where:

HRPS₁₀₀ represents the high reactor pressure setpoint that correspondsto 100% power; and

RP₁₀₀ represents the rated reactor pressure that corresponds to 100%power.

As can be seen, PDHRPS may include a zero-order (constant) component(control value+HRPS₁₀₀−RP₁₀₀), a first-order (linear) component(P/P₀*PCB), and/or a second-order (quadratic) component ((P/P₀)²*SLPD).

This equation can be rewritten as:

PDHRPS=operating pressure+HRPS₁₀₀−RP₁₀₀

The PDHRPS may initiate a reactor scram. In addition or in thealternative, the PDHRPS may initiate one or more warnings, alarms, ormitigating actions. The high reactor pressure setpoint that correspondsto at least one value of percent power in an operating domain of thereactor also may initiate a reactor scram. Additionally, the HRPS₁₀₀ mayinitiate a reactor scram.

The PDHRPS initiating a reactor scram may result in earlier pressurescrams to improve thermal limits due to anticipated operationaltransients, particularly slow pressurization transients and/ortransients that do not have a direct scram from valve closure(s).

In another example embodiment, a protection system for a nuclear BWR mayinclude a first high reactor pressure setpoint that corresponds to 100%power and at least one second high reactor pressure setpoint thatcorresponds to one or more values of percent power in an operatingdomain of the reactor. The at least one second high reactor pressuresetpoint may be less than the first high reactor pressure setpoint.

In yet another example embodiment, a method of operating a nuclear BWRmay include implementing, in a protection system for the reactor, aPDHRPS. The high reactor pressure setpoint that corresponds to at leastone value of percent power in an operating domain of the reactor may beless than the high reactor pressure setpoint that corresponds to 100%power.

In a further example embodiment, a method of operating a nuclear BWR mayinclude implementing, in a protection system for the reactor, a firsthigh reactor pressure setpoint that corresponds to 100% power, andimplementing, in the protection system for the reactor, at least onesecond high reactor pressure setpoint that corresponds to one or morevalues of percent power in an operating domain of the reactor. The atleast one second high reactor pressure setpoint may be less than thefirst high reactor pressure setpoint.

FIG. 9 is a graph showing a typical curve of reactor pressure (psia)versus reactor power (% of rated power) for a BWR, a related artpressure setpoint line, and two example NPSCs that may be linear.

The PDHRPS may be calculated as:

PDHRPS=HRPS₁₀₀ −S*P/P ₀

where S represents a slope determined, for example, to optimize themargin between the operating pressure and the PDHRPS through at leastsome portion of the operating domain of the reactor.

FIG. 10 is a graph showing a typical curve of reactor pressure (psia)versus reactor power (% of rated power) for a BWR, a related artpressure setpoint line, and an example NPSC that may be a series ofconstant values, each covering a range of reactor powers, in the mannerof a step function. The ranges may be, for example, of similar ordissimilar span of percent powers. For the range of reactor powers thatincludes 100% reactor power, PDHRPS may equal HRPS₁₀₀.

FIG. 11 is a graph showing a typical curve of reactor pressure (psia)versus reactor power (% of rated power) for a BWR, a related artpressure setpoint line, and an example NPSC that may be a combination ofone or more, for example, zero-order, first-order, second-order,higher-order, geometric, logarithmic, exponential, step function, and/orother components, each covering a similar or dissimilar span of percentpowers. The example NPSC also may include, for example, one or morelook-up tables, series of points, digital approximations, or othercomponents. For the range of reactor powers that includes 100% reactorpower, PDHRPS may equal HRPS₁₀₀.

Considerations regarding the at least one second high reactor pressuresetpoint are similar to those regarding the PDHRPS.

FIG. 12 illustrates a protection system 200 for a nuclear boiling waterreactor according to example embodiments. The protection system 200 mayinclude a device 202 configured to monitor reactor power; a device 204configured to monitor reactor pressure; a device 206 configured todetermine a power-dependent high reactor pressure setpoint based on themonitored reactor power; and a device 208 configured to initiate aprotection system action when the monitored reactor pressure is greaterthan the power-dependent high reactor pressure setpoint.

As discussed above, the RPS monitoring of valve closure(s) may change,for example, below about 25% reactor power, below about 30%-40% reactorpower, below about 40%-60% reactor power, and/or at or near additionalvalues of reactor power related to other plant-specific needs and designfeatures. Changes in the NPSCs may or may not reflect one or more ofthese changes to the RPS monitoring of valve closure(s).

RPSs include, for example, analog, digital, or analog and digitalcomponents. Digital components may allow, for example, employment ofmore complex NPSCs that may provide additional improvements in safetymargin for a given plant.

In addition to the PDHRPS and/or the at least one second high reactorpressure setpoint, one or more signals in the system that correspond toone or more values of percent power may be delayed in time before theone or more signals affect the power-dependent high reactor pressuresetpoint. The delay may result in the power change during apressurization transient not significantly changing the PDHRPS and/orthe at least one second high reactor pressure setpoint during thetransient. This may be particularly true during slow pressurizationtransients.

The one or more signals may be delayed in time by greater than or equalto about 1 second. For example, the one or more signals may be delayedin time by greater than or equal to about 2, 3, 4, 5, 6, 7, 8, 9, 10,11, 12, 13, 14, 15, 16, 17, 18, 19, 20, 25, 30, 45, 60, or more seconds.The one or more signals may be delayed in time by less than or equal toabout 60 seconds. For example, the one or more signals may be delayed intime by less than or equal to about 45, 30, 25, 20, 19, 18, 17, 16, 15,14, 13, 12, 11, 10, 9, 8, 7, 6, 5, or fewer seconds. The one or moresignals may be delayed in time by greater than or equal to about 1second and less than or equal to about 60 seconds. The one or moresignals also may be delayed in time by greater than or equal to about 1,2, 3, 4, 5, 6, 7, 8, 9, 10, or more seconds and less than or equal toabout 20, 19, 18, 17, 16, 15, 14, 13, 12, 11, 10, 9, 8, 7, 6, 5, orfewer seconds. The delay may be optimized, for example, to obtain themost effective system for the most desirable combination of theoperating thermal limit(s) and/or operating flexibility.

In an example embodiment, the one or more signals may be delayed in timeby greater than or equal to about 5 seconds and less than or equal toabout 10 seconds. In another example embodiment, the one or more signalsmay be delayed in time by greater than or equal to about 6 seconds andless than or equal to about 8 seconds. In yet another exampleembodiment, the one or more signals may be delayed in time by greaterthan or equal to about 8 seconds and less than or equal to about 10seconds. Apparatuses and methods to delay in time the one or moresignals in the system that correspond to one or more values of percentpower are known by one of ordinary skill in the art.

In addition or in the alternative, one or more signals in the systemthat correspond to one or more values of percent power may be laggedrelative to one or more signals in the system that correspond to reactorpressure. The lag may result in the power change during a pressurizationtransient not significantly changing the PDHRPS and/or the at least onesecond high reactor pressure setpoint during the transient. This may beparticularly true during slow pressurization transients.

The lag may be implemented, for example, by electronically filtering themeasured neutron flux signal in a manner similar to that used to producethe STP signal. The measured neutron flux signal may be, for example,the Average Power Range Monitor (“APRM”) signal. The lag may beimplemented, for example, by a single-time-constant filter that mayapproximate the loop transit time of the reactor and/or a value that mayenvelop a typical timing for one or more slow pressurization transients.Such loop transit times may be, for example, greater than or equal toabout six seconds and less than or equal to about eight seconds.Apparatuses and methods to lag the one or more signals in the systemthat correspond to one or more values of percent power relative to oneor more signals in the system that correspond to reactor pressure areknown by one of ordinary skill in the art.

While example embodiments have been particularly shown and described, itwill be understood by those of ordinary skill in the art that variouschanges in form and details may be made in the example embodimentswithout departing from the spirit and scope of the present invention asdefined by the following claims.

1. A protection system for a nuclear boiling water reactor, theprotection system comprising: a device configured to monitor reactorpower; a device configured to monitor reactor pressure; a deviceconfigured to determine a power-dependent high reactor pressuresetpoint, based on the monitored reactor power; and a device configuredto initiate a protection system action when the monitored reactorpressure is greater than the power-dependent high reactor pressuresetpoint; wherein the power-dependent high reactor pressure setpointthat corresponds to at least one value of percent power in an operatingdomain of the reactor is less than the power-dependent high reactorpressure setpoint that corresponds to 100% reactor power.
 2. The systemof claim 1, wherein the protection system action is a reactor scram. 3.The system of claim 1, wherein the protection system action is awarning.
 4. The system of claim 1, wherein the protection system actionis an alarm.
 5. The system of claim 1, wherein the protection systemaction includes a reactor scram.
 6. The system of claim 1, wherein theprotection system action includes a warning.
 7. The system of claim 1,wherein the protection system action includes an alarm.
 8. The system ofclaim 1, wherein the device configured to initiate a protection systemaction initiates a reactor scram when the monitored reactor power isgreater than the power-dependent high reactor pressure setpoint thatcorresponds to at least one value of percent power in the operatingdomain of the reactor.
 9. The system of claim 1, wherein the deviceconfigured to initiate a protection system action initiates a reactorscram when the monitored reactor power is greater than thepower-dependent high reactor pressure setpoint that corresponds to 100%reactor power.
 10. The system of claim 1, further comprising: one ormore signals in the system that correspond to one or more values ofpercent reactor power; wherein the one or more signals in the systemthat correspond to one or more values of percent reactor power aredelayed in time before the one or more signals in the system thatcorrespond to one or more values of percent reactor power affect thepower-dependent high reactor pressure setpoint.
 11. The system of claim1, further comprising: one or more signals in the system that correspondto one or more values of percent reactor power; and one or more signalsin the system that correspond to reactor pressure; wherein the one ormore signals in the system that correspond to one or more values ofpercent reactor power are lagged relative to the one or more signals inthe system that correspond to reactor pressure.